Study on the Sensitivity and Uncertainty of Nuclear Data to the Sodium-Cooled Linear Breed-and-Burn Fast Reactor Using SCALE6.2 Code

Previously, the neutronics design of a small and compact linear breed-and-burn fast reactor (B&BR) was completed. The reactor produces 400 MWth power, and it can operate with excess reactivity of less than 1$ for more than 50 years without refuelling. As the blanket fuel, the spent nuclear fuel (SNF) from existing light water reactors (LWRs) is used to reduce the burden from the problematic long-lived isotopes in SNF. However, by loading massive nuclides at the initial core, the impact of nuclear data uncertainty on the reactivity calculation results of SNF-fuelled B&BR at the beginning of life (BOL) is expected to be significant because these nuclides have different credentials in evaluated nuclear data libraries. In this study, the impact of nuclear library uncertainty from ENDF/B-VII.0 and ENDF/B-VII.1 on reactivity calculation of B&BR is evaluated using the continuous-energy TSUNAMI-3D module in the SCALE6.2 code package. The uncertainty of reactivity calculation results of B&BR caused by the inaccuracy of two libraries is significant (more than 2000 pcm), mainly from the uncertainty of 235,238U and 56Fe cross section. The energy-dependent sensitivity profiles show that they are significant at the fast energy range. The uncertainty of coolant void reactivity (CVR) is about 18%, and that of fuel temperature coefficient (FTC) is about 15% of the reactivity effect. The top five contributions for CVR accounted for elastic scattering of 238U, capture of 235,238U, and elastic scattering of 23Na and 56Fe. Meanwhile, the top contributors for FTC were accounted for elastic scattering of 238U and 56Fe, capture of 235U, and elastic scattering of 94Zr and 57Fe. It is highly recommended to improve the accuracy of those isotopes’ cross sections at the high energy range to provide a more reliable reactivity calculation for the fast system.

FUEL TAP: A SIMPLIFIED BREED AND BURN MSR

Breed-and-burn Molten Salt Reactors are an interesting option of reactor design that allow high fuel utilization while operating on an open fuel cycle. Such reactors usually require specialized codes in order to model its fuel cycle and the flowing fuel in an unmoderated core. In this work, we propose a design and perform a preliminary analysis of a homogeneous chloride salt single-fluid design. The fuel cycle is analyzed using the EQL0D tool in order to model reactor start-up and transition into an equilibrium state. Core simulation is performed using ATARI, an OpenFOAM-based multiphysics code developed at PSI. Results show that the core size for such a reactor is quite big and that it can be easily started with high-assay LEU. In addition, the core has been designed to promote a quasi-1D flow, opening the possibility of modeling the core with legacy codes in the future.

SIMULATION OF BREED AND BURN FUEL CYCLE OPERATION OF MOLTEN SALT REACTOR IN BATCH-WISE REFUELING MODE

The Molten Salt Reactor (MSR) is one of the most revolutionary Gen-IV reactors and it can be operated, especially with chloride salts, in the so-called breed and burn fuel cycle. In this type of fuel cycle the fissile isotopes from spent fuel do not need to be reprocessed, because the excess bred fuel covers the losses. The liquid phase of the MSR fuel assures its instant homogenization, and the reactor can be operated with batch-wise refueling thus reaching an equilibrium state. At the same time, the active core of the chloride fast MSR needs to be bulky to limit neutron leakage. In this study, the code Serpent 2 was coupled to the Python script BBP to simulate batch-wise operation of the breed and burn MSR fuel cycle. The script, previously developed for solid assemblies shuffling, was modified to simulate fuel homogenization after fertile material addition. Several fuel salts and fission products removal strategies were simulated and their impact was analyzed. Similarly, the influence of blanket volume was assessed in a two-fluid core layout. The results showed that the reactivity initially grows during the irradiation period and later decreases. The blanket has a large impact on the performance and it can be used to further increase the fuel burnup or to shrink the active core size. The breed and burn fuel cycle in MSR can reach high fuel utilization without fuel reprocessing and a multi-fluid layout can help to decrease the core size.

CORE DESIGN OF BREED & BURN MOLTEN SALT FAST REACTOR

Previous designs of once-through solid-fuelled breed-and-burn (B&B) reactor and the conventional molten salt reactor (MSR) concepts suffer from material limitation of neutron irradiation damage and chemical corrosion. A novel breed-and-burn molten salt reactor (BBMSR) concept uses separate molten salt fuel and coolant in a linear assembly core configuration. Similar to Moltex Energy Stable Salt Reactor (SSR) design, the configuration with fuel salt contained in fuel tubes and coolant salt in pool type reactor vessel has been previously studied. The study confirmed that breed-and-burn operation is feasible in principle, however with a low neutronic margin. The objective of this paper was to seek improvements of the neutronic margin with a metallic natural uranium blanket design. A parametric study was performed for the natural uranium blanket design. BBMSR neutronic performance simulation was modelled using Serpent, a Monte Carlo reactor physics code, with a single 3D hexagonal channel containing a single fuel tube in an infinite lattice with reflective radial and vacuum axial boundary conditions. The addition of a metallic natural uranium blanket inside the fuel tube, which increases the natural uranium metal to fuel salt ratio (ϒ) of the BBMSR, was shown to significantly increase the neutronic performance of the BBMSR.